The NJOY nuclear data processing system is a comprehensive computer code system for producing pointwise and multigroup cross sections and related quantities from ENDF/B evaluated nuclear data in the ENDF format, including the latest US library, ENDF/B-VI. The NJOY code works with neutrons, photons, and charged particles and produces libraries for a wide variety of particle transport and reactor analysis codes.

It is capable of processing data in ENDF/B-4, ENDF/B-5, and ENDF/B-6 formats for evaluated data (to the extent that the latter have been frozen at the time of this release).

NJOY94.61 consists of a set of modules, each performing a well-defined processing task. Each of these modules is essentially a separate computer program linked together by input and output files and a few common constants. The methods and instructions on how to use them are documented in the LA-12740-M report on NJOY91 and in the README file. No published document is planned for NJOY94, which is an interim release on the way to NJOY95. NJOY94.0 included all the changes in NJOY91.118, the up119 patch, the PostScript plotting package, the new LEAPR module for computing thermal scattering laws, and a major update to PURR incorporating a new version of the unresolved resonance probability table method including temperature correlations. The November 1995 update to RSIC's package was to replace NJOY94.5 which was packaged in June 1995 with NJOY94.10, which was used to process all of the materials from ENDF/B-VI Release 3 into PENDF format, ACE format, and a 69-group library. In the course of this work, a few additional problems were found and corrected by updates 6-10, which involve changes to the therm, error, acer and njoy modules. Details on the August 1996 NJOY94.35 can be found in the README35 file. In December 1996 the package was upgraded to NJOY94.61.

RECONR:
Reads an ENDF/B tape and produces a common energy grid for all reactions (the union grid) such that all cross sections can be obtained to within a specified tolerance by linear interpolation. Resonance cross sections are calculated with the methods of RESEND (17) but a new method of choosing the energy grid is used which incorporates control of the number of significant figures generated and a resonance-integral criterion to reduce the number of grid points generated for some materials. Summation cross sections (e.g. total, inelastic) are reconstructed from their parts. The resulting pointwise cross sections are written onto a ?point-ENDF" (PENDF) tape for future use.

BROADR:
Reads a PENDF tape and Doppler broadens the data using the method of SIGMA1 (18), modified for better behaviour at high temperatures and low energies. The union grid allows all resonance reactions to be broadened simultaneously resulting in great savings of processing time. After broadening, the summation cross sections are again reconstructed from their parts. The results are written out on a PENDF tape for future use.

UNRESR:
Uses the methods of ETOX (19) to produce effective self-shielded pointwise cross sections, versus temperature and background cross section, in the unresolved-resonance region. The results are added to the PENDF tape in a special format.

HEATR:
Computes both heating and radiation-damage-energy production using momentum balance (for capture) or energy balance (for all other reactions). The ENDF/B photon production files are used in both methods, when available. The heating results are added to the PENDF tape using ENDF/B reaction numbers in the 300 series, and the radiation damage results are the special identifier 444.

THERMR:
Produces cross sections in the thermal range. Bragg edges in coherent scattering are produced using the method of HEXSCAT (16) with an improved treatment at high energies.
Energy-to-energy incoherent scattering matrices can be computed for free scattering or for bound scattering using a precomputed form factor S(alpha,beta) in ENDF format. The secondary angle and energy grids are determined adaptively so as to represent the function to a desired precision by linear interpolation; the angular representation is converted to one based on equally-probable angles. Elastic incoherent scattering is represented using equally-probable angles computed analytically. The results for all the processes are added to the PENDF tape using special formats and reaction numbers.

GROUPR:
Processes the pointwise cross sections produced by the modules described above into multigroup form using the Bondarenko flux weighting model (20). As an option, a pointwise flux solution can be generated for a heavy absorber in a light moderator. Self-shielded cross sections, scattering matrices, and photon production matrices are all averaged in a unified way, the only difference being in the function which describes the feed" into secondary group g' with Legendre order l from initial energy E. The feed for two-body scattering is computed using a centre-of-mass Gaussian integration scheme which provides high accuracy even for small Legendre components of the scattering matrix. Special features are included for delayed neutrons, the coupled angle and energy dependence of the thermal scattering matrix, and the discrete scattering angles arising for thermal coherent reactions. Prompt fission is treated with a group-to-group matrix. The results are written in a special "groupwise-ENDF" format (GENDF) for later use by the output formatting modules.

GAMINR:
Uses a specialized version of GROUPR. Coherent and incoherent form factors (21) are processed in order to extend the useful range of the results to lower energies. Photon heat production cross sections are also generated. The results are saved on a GENDF tape.

ERRORR:
Can either produce its own multigroup cross sections using the methods of GROUPR or start from a pre-computed set. The cross sections and ENDF covariance data are combined in a way which includes the effects of deriving one cross section from several others. Special features are included to process covariances for data given as resonance parameters or ratios (e.g. fission nu-bar).

COVR:
Uses the widely available DISSPLA (22) plotting software to make publication-quality plots (23) of covariance data; it also provides a site for user-supplied routines to prepare covariance libraries for various sensitivity systems.

DTFR:
Is a simplereformatting code which produces cross section tables acceptable to most discrete-ordinates codes. It also converts the GROUPR fission matrix to chi (fission spectrum) and nu-sigma-fission and prepares a photon production matrix if desired. The user can define edit cross sections which are any linear combination of the cross sections on the GENDF tape. This makes complex edits such as gas production possible. DTFR also contains system-dependent plotting routines for the cross sections and P0 scattering and photon production matrices.

CCCCR:
CCCCR is also a straightforward reformatting code. All of the CCCC-IV (7) options are supported. In the cross section file (ISOTXS), the user can choose either isotope chi matrices or isotope chi vectors collapsed during any specified flux. The BRKOXS file includes self-shielding factors for elastic removal. It should be noted that some of the cross sections producible with NJOY are not defined in the CCCC-IV files.

MATXSR:
Reformats GENDF data into the MATXS file format, which is suitable for input to the TRANSX post-processor program. The MATXS format uses flexible naming conventions which allow it to store all NJOY data types except delayed neutron and delayed photon spectra.

ACER:
Prepares a data library tape for MCN, the LASL continuous energy Monte-Carlo code. Reaction cross sections are written out on the grid of the total cross section from the input PENDF tape (assumed to be linearized and unionized). Redundant reactions are removed. Angular distributions (MF4) and tabulated energy distributions (MF5, LF1 or 5) are converted into equal probability bins. The incident energy grids of both are thinned for linear interpolation using the specified tolerance. Analytic secondary energy distributions are copied. All photon production cross sections are combined on the cross section energy grid and written as MF13,MT3. Multigroup photon production cross sections are obtainedfrom the NGEND input tape.

The photon distributions are summed and converted into a set of equally probable mean energies and written as MF15,MT3LF3 (a specially defined law). MF14 is made isotropic. Unresolved region probability tables are generated by a least square fit to the self-shielded cross sections on MT152 (see UNRESR) and written out at MT153 in a special format. The dictionary on the output tape is corrected to reflect the changes.

A Fortran 77 compiler is required. This system runs on Cray under either UNICOS or CTSS, on the VAX under VMS, on the Sun under either Sun OS4 or OS5, IBM RS/6000 under AIX and on DEC Alpha under OSF/1. The CFT77 compiler was used under UNICOS. RSIC tested this release on a
SunSparc 5 running SUN OS 5.3 using the Fortran 2.0 compiler and on the IBM RS/6000 Model 590 under AIX 3.2.5 using the XLF 3.2.2.3 compiler. NJOY 94 produces graphs directly in Postscript and no longer requires the proprietary DISSPLA software.

PSR-0355/02

The test compilation and executions were done at the NEA Data Bank under OSF/1 using the standard Fortran compiler.